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At least 865 records · Page 48
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Design of a High-Assay Low-Enriched Uranium Tri-Structural Isotropic Critical Experiment for Advanced Reactor Validation

High-assay low-enriched uranium (HALEU) fuel is a key component of many small modular reactor designs. Critical experiments are an important way to understand the neutronic performance of systems by obtaining nuclear data validations through measurements. Data reduce uncertainty and risk by showing that systems respond as predicted to changes such as temperature, subsequently advancing the overall technology readiness level of the materials within. Numerous critical experiments have been performed at the National Criticality Experiments Research Center (NCERC) operated by Los Alamos National Laboratory at the Nevada National Security Site since it became operational in 2011. However, the first experiment with HALEU fuel did not occur until 2024. Through extensive engineering, the experiment described in this paper was successfully designed and executed for the Comet vertical lift assembly at NCERC to perform measurements with HALEU tri-structural isotropic fuel that will assist in validation of nuclear data and computational modeling of small modular reactors for years to come.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Horizontal Split Table Conceptual Design for Advanced Reactor Validation

Oak Ridge National Laboratory and Lawrence Livermore National Laboratory are collaborating to develop a conceptual design for a horizontal split table for use in performing critical experiments. The goal of this design effort is to provide nuclear data testing and validation capabilities for advanced reactors such as pebble-bed high-temperature gas-cooled reactors, molten salt reactors, and heat pipe microreactor, but it could also be used for the current generation of reactors. The first concept being explored for the horizontal split table, a pebble-bed design based on the HTR-10 reactor, is described in this paper. A critical configuration corresponding to a footprint of about 4.5 m 2 was determined with SCALE/KENO-VI to fit the planned dimensions of the horizontal split table. The similarity of the pebble-bed design and the HTR-10 reactor application was assessed using SCALE/TSUNAMI, and a similarity coefficient $c_k$ of 0.9982 was obtained, proving that the concept will be useful for nuclear data validation and assimilation of pebble-bed type advanced reactors. In the proposed design, the materials with the highest $k_{eff}$ sensitivity are graphite and uranium, demonstrating that particular care must be given to carbon-related cross-section data. The effect of mechanical uncertainties between the fixed and moving tables was also assessed by calculating the reactivity change caused by vertical and horizontal gaps, as well as angular and torsion offsets between the two sides of the horizontal split table concept. The highest relative changes on the concept’s reactivity were caused by angular perturbations. The same analysis process is currently being used to create a molten salt advanced reactor type horizontal split table concept based on the Molten Salt Reactor Experiment (MSRE).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Analytic Sensitivity Coefficients for General Multigroup Infinite Medium k-Eigenvalue Problems

The general multigroup infinite medium k-eigenvalue neutron transport equation is used to derive analytic expressions for the infinite medium k-eigenvalue, the scalar neutron flux and adjoint, and the sensitivity of $k$ ∞ to perturbations in the multigroup nuclear data of a single species isotropic elastic scattering material. In the appendix, we present the multigroup nuclear data for U-235 and U-238 along with the corresponding k-eigenvalue, flux, adjoint, and sensitivity profiles, which include the sensitivity of $k$ ∞ to the total, fission, capture, and scattering macroscopic cross sections as well as to the group-to-group scattering cross section matrix, group neutron production, and the unconstrained and constrained fission neutron energy distribution.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

ENDF/B-VIII.1 Validation (Preliminary Release 2 + New 239 Pu)

This document should serve as a reference for validation testing of nuclear data files included in the second preliminary release of ENDF/B-VIII.1 and the newly evaluated 239 Pu nuclear data file to be included in the third preliminary release of ENDF/B-VIII.1. Experiment names listed in this report are based on International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook designations. Additional information on ICSBEP handbook designations can be found at https://www.oecd-nea.org/jcms/pl_20291/icsbep-handbook. The validation metrics used in this report include the effective neutron multiplication factor, k eff , and tritium production. All validation metrics were calculated using the radiation transport code Monte Carlo N-Particle (MCNP).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Technical note on 16 O(n,α) evaluation (Rev. 3)

The evaluated 16 O(n,α) cross sections in the ENDF/B-VIII.0 nuclear data library remain uncertain because of systematic discrepancies between measured data. The evaluation relies heavily on the inverse reaction, and, in particular, it relies on the recently measured 13 C(α,n) 16 O cross sections reported by Harissopulos in 2005. These cross section values are systematically lower than those previously reported by Bair in 1973. This technical note briefly describes and shows, on the basis of the available experimental information, that the two sets of measured cross sections are consistent

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Extension of SCALE/Sampler’s sensitivity index for the assessment of cross section, fission yield, and decay data uncertainties

Accurate prediction of nuclide compositions and neutron and gamma sources and spectra for fresh and spent nuclear fuel through computational modeling and simulations is the basis for safeguards instruments design, optimization, and calibration, and for validation of the measured responses. The accuracy of these simulations can be significantly impacted by uncertainties in input parameters to the applied computer code. Input parameters are, for example, dimensions, material compositions, and temperatures of the model. Additionally, important but sometimes neglected input parameters for simulations of nuclear systems are the nuclear data that include nuclear reaction cross sections, fission product yields, and decay data. It has been shown that uncertainties of calculated results are dominated by uncertainties of these nuclear data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Improved Cumulative Fission Yield Measurements with Fission Spectrum Neutrons on 235 U

Two isotopically-characterized targets of high-purity (99.84±0.03%) 235 U-oxide and a neutron dosimetry package were exposed to a pulsed neutron irradiation using the Godiva critical assembly at the National Criticality Experiments Research Center. Following irradiation, the fissioned targets were counted with a broad-energy germanium detector. Estimates of the total fissions for each target were made using measurements of well-known fission products. The total fission values were then used to measure the fission yields of 21 fission products with half-lives between 636 and 119,700 seconds. Mean values of seventeen of the measured fission products fell within the 95% confidence level of the Evaluated Nuclear Data Files VII.1/b (ENDF/B-VII.1) using the larger of the two uncertainties: 84g Br, 89 Rb, 91 Sr, 93 Y, 97 Zr, 101 Mo, 105 Ru, 128 Sn, 129 Sb, 130g Sb, 131 Sb, 131m Te, 133 I, 135 I, 141 Ba, 142 Ba, and 142 La. Seven isotopes were measured to better precision than is listed in the ENDF/B-VII.1 database: 89 Rb, 93 Y, 128 Sn, 129 Sb, 130g Sb, 131m Te, and 133 I. Four of the measured isotopes were found to be statistically different from the ENDF/B-VII.1 database: 94 Y, 134 Te, 134 I, and 138 Xe. Two of the isotopic analyses conducted in this paper fit the parent and daughter data simultaneously. The parent-daughter pairs measured in this way were the 134 Te/I and 142 Ba/La pairs. The fission yield results obtained in this way are unique, especially for 134 I, because the parent and daughter half-lives are so close. Here, given the limited number of direct measurements of the 134 Te/I pair from fast neutron induced fission and the seemingly high estimate for 134 Te from previous work, it is likely that the value in this paper is more accurate; however, additional measurements of this pair are needed to validate this assertion.

42 ENGINEERING↗
DOE OSTI2024

Chlorine Validation through Critical Integral Experiments

An increased interest in the validation of chlorine nuclear data, specifically 35Cl, in recent years has led to the prioritization of conducting chlorine critical experiments for nuclear data validation and criticality safety evaluations. A previous experiment was conducted with PVC and CPVC (chlorinated polyvinyl chloride) to reduce the margin of subcriticality for PF-4 operations with aqueous plutonium chloride solutions. Challenges in determining the composition of specific polymers, namely the CPVC, made it difficult to characterize, and therefore benchmark. Additional materials that were utilizable for chlorine validation were not immediately clear, as form, strength, purity, and composition are all important. A series of possible absorber materials were identified and studied, but ultimately granular sodium chloride was chosen. In addition to the needs of LANL’s PF-4 other industry collaborators have brought attention to the need for chlorine validation. Specifically, the need for HEU electrorefining with LiCl salts at Y-12 was taken into consideration for this experiment, with the experimental configurations being finely tuned to best meet their needs. Needs for validation were also presented by Idaho National Lab (INL) and TerraPower for validation of their Molten Chloride Reactor Experiment (MCRE) and Molten Chloride Fast Reactor (MCFR) and have also been considered.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Monte Carlo Analyses of the FOEHN Experiment

The FOEHN critical experiment, which has been carried out at Cadarache (France) in the reactor EOLE in the early 70s, was designed to verify RHF design analyses. This study focuses on the impact of the FOEHN experiment uncertainties on the calculation of the energy deposition. The latter has been obtained by the Monte Carlo codes Serpent and MCNP. In the calculation of the energy deposition, the first code relies on KERMA factors whereas the latter code relies on Q-values. The two Monte Carlo codes share the same geometry and material specifications. In addition to the design and experiment parameters uncertainties, nuclear data also impacts the calculation of the energy deposition. In this study, the MCNP simulations use three different nuclear data library sets: ENDF/B versions VII.0, VII.1, and VIII.0. Finally, the ksens card of MCNP, using perturbation theory, has been used to identify the cross sections and isotopes with largest impact on the effective multiplication factor. By showing that the use of different software and computational methodologies and design parameters uncertainties do not significantly impact the core power distribution, this work demonstrate the appropriateness of the tools and methods employed for the conversion of involute-plate reactors.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Radiochemical separations and experimental measurements of short-lived fission products from 14 MeV irradiation of depleted uranium

Here, this work focused on the experimental analysis of depleted uranium irradiated with 14 MeV neutrons to obtain fission yields for 20 fission products with half-lives ranging from 16.8 h to 30.1 years. Several analytes were identified where improvements to the available nuclear data would be beneficial. The results obtained from this experiment could be used to make improvements to cumulative fission yield data once confirmatory measurements are made. The required separations will be described along with a comparison to the reported Evaluated Nuclear Data File fission yields and other experimental measurements.

14 MeV fission↗
DOE OSTI2021

Sensitivity-based Experiment Design Optimization for a Molybdenum Critical Experiment

A lack of intermediate molybdenum benchmarks in the ICSBEP has been identified by LANL, Y-12, and IRSN. This lacking adversely effects criticality safety operations and leaves new differential molybdenum data unvalidated. NCERC is proposing a series of intermediate integral experiments to better the understanding of molybdenum systems. Using MCNP6.2 with the ENDF/B-VIII.0 nuclear data library a single unmoderated and four moderated system designs were identified using a sensitivity optimization method. Each proposed system was found to be at least twice as sensitive to the 95 Mo capture cross section in the URR as the sole existing intermediate molybdenum benchmark in the ISCBEP handbook. The addition of a new molybdenum sensitive intermediate system would improve future nuclear data evaluations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

SCALE Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations

Sensitivity coefficients describe the fractional change in a system response that is induced by changes to system parameters and nuclear data. The Tools for Sensitivity and UNcertainty Analysis Methodology Implementation (TSUNAMI) code within the SCALE code system makes use of eigenvalue sensitivity coefficients for an extensive number of criticality safety applications, including quantifying the data-induced uncertainty in the eigenvalue of critical systems, assessing the neutronic similarity between different critical systems, and guiding nuclear data adjustment studies. The need to model geometrically complex systems with improved fidelity and the desire to extend TSUNAMI analysis to advanced applications has motivated the development of a methodology for calculating sensitivity coefficients in continuous-energy (CE) Monte Carlo applications. The Contributon-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) and Iterated Fission Probability (IFP) eigenvalue sensitivity methods were recently implemented in the CE-KENO framework of the SCALE code system to enable TSUNAMI-3D to perform eigenvalue sensitivity calculations using continuous-energy Monte Carlo methods. This work provides a detailed description of the theory behind the CLUTCH method and describes in detail its implementation. This work explores the improvements in eigenvalue sensitivity coefficient accuracy that can be gained through the use of continuous-energy sensitivity methods and also compares several sensitivity methods in terms of computational efficiency and memory requirements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

(U) Average Scattering Angle Cosine Sensitivities

The Evaluated Nuclear Data File (ENDF) provides covariances of the average scattering angle cosine $\overline{μ}$ for 117 nuclides, and the NJOY nuclear data processing code can process these covariances. To propagate them to the uncertainty of a response due to uncertainties in $\overline{μ}$ requires sensitivities of the response with respect to $\overline{μ}$. This report uses NJOY’s definition of $\overline{μ}$ and derives a relationship between the sensitivity of a response to $\overline{μ}$ and the sensitivity of the response to the anisotropic P 1 elastic scattering cross section. This relationship has been implemented in the multigroup neutron sensitivity code SENSMG, which relies on the discrete ordinates code PARTISN for the neutron transport calculations. Kodeli has also calculated sensitivities with respect to $\overline{μ}$ and associated uncertainties. However, the formal derivation of this paper shows that the definition of the sensitivity with respect to $\overline{μ}$ is ambiguous because there are multiple ways to change $\overline{μ}$ by changing scattering cross sections. Here we derive two such ways. Both are implemented in SENSMG. This report is organized as follows. Section II discusses $\overline{μ}$ and its significance. Section III presents the equations used for the sensitivity of a response with respect to $\overline{μ}$, and Section IV discusses how to compute these sensitivities with SENSMG and MCNP6.3. Section V is a fast-spectrum test problem, and Sec. VI is a thermal-spectrum problem. Section VII is a summary. The example problem inputs are listed in the appendix.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Understanding Peelle’s Pertinent Puzzle bias in generalized least squares regression through eigenspectrum analysis

Certain correlation structures in the data covariance matrix (DCM) used for generalized least squares (GLS) regression can result in biased estimates, commonly known in the field of nuclear data evaluation as Peele’s Pertinent Puzzle (PPP). This article introduces a generative, forward modeling framework within which the PPP bias is characterized through an eigenspectrum analysis of the DCM. This analysis highlights the root cause of the bias, generalizes the problem beyond the nuclear data field, and provides insight to the problem regimes where it can occur. What follows is an understanding that the bias can show up for any experimental neutron time-of-flight data for which systematic uncertainties have been quantified. Lastly, a discussion of the adaptation of cross validation approaches that require pre-whitening to incorporate the known ‘fix’ to the PPP bias in the GLS estimator.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Nuclear Criticality Safety Integral Experiment Covariance Determination

Integral benchmarks for criticality safety and nuclear data validation require expensive uncertainty quantification studies. Commonly, the uncertainty quantification ignores correlations between experiments that share components. Experiments such as the TEX (Thermal/Epithermal eXperiments) campaigns consist of many shared parts between experiments, such as fuel, which creates a strong correlation in their errors. While these correlations are known to exist, they are often not estimated due to the complexity of such calculations. This paper describes a software package that uses an intuitive method of determining the covariance for each of the experimental components, providing a correlation matrix for each family of parts across the multiple cases examined within a benchmark. The code uses the TEX-HEU campaign as a proof of concept, and we show that the correlations can be calculated with information commonly found in ICSBEP (International Criticality Safety Benchmark Evaluation Project) benchmarks. The estimated covariances are used in χ 2 trending studies to evaluate their impact on nuclear data validation. The covariance determination code can be easily integrated into current benchmark evaluations as well as reevaluating legacy benchmark uncertainties. Uncertainty correlation calculations should become the baseline for criticality safety integral experiment benchmarks and can now be easily calculated with the described software package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Activation product interpretation of structural material for fast critical assemblies

This work evaluates whether activation products in stainless steel components can be used to infer initial conditions (i.e. alloy composition, position within the fluence, etc.) and the associated uncertainties in a fast neutron irradiation. Activation measurements of stainless steel and elemental components were obtained through a Flattop critical assembly irradiation experiment. A forward model was developed, utilizing MCNP6.1 and FISPACT ii, to estimate activation product concentrations from several input parameters: alloy composition, source specifications, and foil positioning. While some of the calculated activation product concentrations agreed fairly well with measurements, some isotopics (i.e. 57 Co) had large discrepancies. It became clear that uncertainties in both measurements and nuclear data drove this divergence for select isotopics, and the quantification of this impact was needed. By treating both experimental measurement and nuclear data uncertainty in a Bayesian framework, it was possible to infer unknowns from given activation ratios. The developed approach showed promise in constraining mass parameters, but degeneracy existed when considering position within the neutron flux. Higher threshold reactions could potentially break the degeneracy, such as 57 Co, but uncertainty estimation needs to be better resolved for those isotopes.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Charged Particle Transport Libraries (Final Report FY2020 Level-2 Milestone #7127)

This document is the final report on the status and developments of the Charged-Particle Transport (CPT) Level-2 (“CP2020”) ASC/PEM Milestone work. This section gives an overview of the CP2020 work partitioned into two subsections: 1) a summary of the code tools used for generating CP (and other) data in Section 1.2; and 2) and library data and codes access information for end users in Section 1.3. Section 2 gives an overview of the status of previous work on charged-particle (CP) nuclear data, particularly focused on the work of the FY2011 CP Level-2 Milestone. New work for CP2020 is detailed in the following Section 3. Section 4 discusses testing and validation of the evaluated and processed data for the CP2020 library. Conclusions based on the experience of this Level-2 work and the outlook for future development of CPT-relevant data are given in the final Section 5. Appendices, detailed throughout the report, are provided for further information. This report, related documentation, codes, and the CPT nuclear data libraries are available via online repository.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗